Method of recovering transuranic elements of an atomic number below 95



Dec. 15, 1959 G. T. sx-:ABoRG :TAL 2,917,361

METHOD 0F' RECOVERING TRANSURANIC ELEMENTS OF AN ATOMIC NUMBER BELOW 95 f Filed Feb. 11, 194s METHOD F RECOVERING TRANSURANIC ELE- MENTS OF AN ATMIC NUMBER BELOW 95 Glenn T. Seaborg and Ralph A. `lames, Chicago, Ill.,

assignors to the United States of America as represented by the United States Atomic Energy Commission Application February 11, 1946, Serial No. 646,946 '6 Claims. (Cl. 23-14.5)

This invention relates to the concentration Vof transuranic elements contained in dilute solutions thereof Yor in mixtures with larger amounts of other metal compounds. The invention is especially concerned with the concentration of compounds of element 93 and element 94 contained in less than millimolar concentrations in solutions derived from lneutron-irradiated uranium.

An object of the present invention is to provide a process for concentrating elements 93 and 94 in solutions so dilute as to make impractical the direct recovery of a precipitate consisting solely of insoluble compounds of elements 93 and 94. l

Another object Yof this invention is to provide a multistage process for concentrating element 94 in such solutions to an extent sufcient toenablethe recovery of a nal precipitate of a substantially pure compound of ele.- ment '94.

"A further object of the present invention is to provide a method for concentrating element 94 in mixtures containing ions of element 9,4 and cations of coprecipitable carrier compounds, whereby the ratio of carrier cation to element 94 may be substantially reduced.

"Additional objects and advantages of this invention will be evident from the following description, and the accompanying drawing which sets forth diagrammatically the Yprincipal steps of a preferred modiiic'ation ofthe inventionf The term element 94 isused throughout this specitication to designate the element having atomic number 94. The designation94239 refers to the isotope ofelement 94 having a mass numberrof 239. Element 94 is also referred to in this specification, and'probablyY will become known in the art, as plutonium, symbol Pu. Likewise, element 93 means the'e'lement'having atomic number 93, which is also referred to as neptunium, Vsymbol Np. Reference herein to any of the elements is to be understood as denoting the element generically, whether in its free state, or in the form of a compound, unless otherwise indicated by the context.

The apparent discovery of transuranic elements (elel ment 93 and elements of higher number) was lirst anmounced by E. Fermi in 1934. AAt that time YFermi stated that the bombardment of uranium with neutrons vgave beta activities which he Y attributed to transuranio elements of atomic number 93 and possibly higher.

From 1934 to 1938 other workers,gnotably Hahn and Curie, extended this work. But in 1939, Hahn discovered that the elements which he and others had believed to'be transuranic elements werein fact radioactive elements of medium atomic weight producedr by the lission of uranium. Hahns results'were subsequently confirmed, and a great many fission products'in addition to those first found by Hahn were discovered andri'dentitied. Such products were all Vof lower atomic number than uranium, generally of atomic numbers in the middle of the periodic group; and so far as is now known, prior to about June 1940, no evidence was found to establish the existence of'any transuranic element.

However, in June 1940, McMillan and Abetson pub- 2,917,361' l Patented Dec. 15,1955) lished in -The Physical Review, 57, 1185, their discovery thatra 2.3-day activity produced by the bombardment of uranium with neutrons was an isotope of element 93, probably ,93239. Although McMillan Yand Abelson surmised that element 94 would be formed by beta. decay of element 9323?, they were unable to produce any evidence of its existence, and did notobtain either 93239 or 9423,9.in pure form or in microscopic amounts, either as the element or'as acompourid. y

E. Segre, G. T. Seaborg and J. W. Kennedy, using ther methods of McMillan and Abelson, obtained 93239admixed with rareV earths, proved that 93,239 decayed to 94239, and measured the radioactive and fission properties lof 94239. Subsequently, neutronic reactors were developed for, the production of 93239 and 94239 in isolable quantities by a self-sustaining chain reaction of slow neutrons with U235 and U238 in natural uranium. 'Y Y Natural uranium comprises largely isotope Um, together with about 1/139 as much H2735 and a very much smaller amount of Um. When this mixture of isotopes, either as metallic .uranium or as a uraniumcompound, is subjected to bombardment by neutrons from an external source, or undergoes a self-sustaining neutronic chain reaction, a number of nuclear reactions take place. Isotope U238 captures a neutron to form U23", which undergoes vbeta decay yto form transuranic elements as shown by the following equation: Y

23mm. (Bo

Uzas oni Um 94239 'cessive steps, leading ultimately to stable isotopes of higher nuclear charge thanthe original fragments.` 'Ihe fission of U235 is predominantly binary, and may be exempliecl by the following type of equation: Y

Substantially all of the fission fragments have mass numbers within the range 77,-158, although small quantities of isotopes of lower and higher mass numbers may result from unbalanced binaryY fissions, ternary hssons, or other reactions of infrequent occurrence. A very large majority of the'fission fragments comprises a light group of mass numbers 8 4-106 andV a heavy group of mass numbers 128-l5Q.

The various decay products ofk the initial iission fragment are referred to herein as fission products.Y These fission products fall within a range of atomicnurnbegs from about 32 torabout 64. The fission products from the light group of fragments referred` to above have atomic numbers ranging from about v32- to kabout 46:;

and the fission products from the heavy group of fragments have atomic numbers ranging from about 5l to about 64. I y The various radioactive Viission Vproducts have, halflives ranging from a fraction of a second to thousands of years. Those having very short half-lives may be eliminated by aging the material for a reasonable period before handling. Those with very long half-lives do, not have sufficiently intense radiation to endanger personnel protected by moderate shielding. On the other hand,V the fission products having half-lives ranging .from a few days 'to a few yearsl have dangerously intense radiations which cannot be eliminated by aging for practical kstorage pe.- riods. These products are chiefly radioactive isotopes of Sr, Y, Zr, Cb and Ru of the light group andY Te, l, Cs, Ba, La, Ce, and Pr of the heavy group.A

U238- innatural uranium by neutron lbombardment is a function of neutron density and of the time of bombard ment. Since 94239 is ssionable under the conditions for iission of U235, the net yield of 94239 per unit of time will decreaseas the 'ratio of U235 content to 94239'content `of the mass decreases. For this reason, a neutronic reaction of 94239 production is suitably terminated when only afraction of the U23-5 has been converted to fission products, The reaction mass at this point contains a largefamount of U23, a much smaller amount of U235, still smaller amounts of 93239, 9,4239 and iission products, and traces of, other products such as UXl and UX2. By aging such a mass `for a suitable period of time, the 93239 may be substantially completely converted to 94239, `with simultaneous conversion of the short-lived radioactive fission products to longer-lived or stable isotopes.

The plutonium content of the aged reaction mass will usually b'econsiderably less than 1 percent of the Weight `of the unreacted uranium, and may even be less than one part per. million parts of uranium.l The concentration of fission products will be of the same order of magnitude. The separation of plutonium from such a mass involves extraction from the unreacted uranium, decontamination by separating the radioactive iission products, and concentration of the decontaminated material to obtain a product for which a relatively pure plutonium compound can be directly recovered.

Techniques have been developed for extracting plutonium from a solutionl of neutron-bombarded uranium by means of a carrier precipitate such as bismuth phosphate or lanthanum fluoride. The precipitates thus obtained are substantially free from uranium but contain sufficient fission products to have intense radiations. Similar techniques have also been developed for the decontamination of'the extracted plutonium by means of a carrier precipitate. Such methods usually involve alternately carrying plutonium in one state of oxidation with the precipitate, and leaving plutonium in solution in a second state of oxidation while carrying fission prodp ucts with the precipitate. After Adecontamination by Asuch methods, the plutonium `may be substantially free from dangerously radioactive fission products, but it is still associated in relatively minute amounts with large quantities of a carrier precipitate or with large volumes of solution containing ions of the preceding carrier precipitate. e

Solutions of plutonium obtained by the foregoing procedute have such low plutonium concentrations that the direct precipitation of even the most insoluble plutonium compounds is either impossible or wholly impractical. The application of common concentration procedures to such solutions results in concentration of carrier ions as well as plutonium ions, and increases the diiiiculties of subsequent plutonium separation.

In accordance with the present invention, the plutonium insuch solutions may be concentrated by successive use of a carrier, with decreasing ratios of carrier to plutonium, employing carriers having certain characteristics hereinafter set forth which make possible this method of operation. The term carrier as used `herein and in the appended claims signifies a substantially insoluble, solid, finely divided compound capable of ionizing to yield at least one inorganic cation and to yield at least one anion which constitutes an ionic component of a compound containing the ion to be carried and which is not substantially more soluble than said finely divided compound in the solution from which carrying is to be elected. A large number of compounds having these essential characteristics are known to be carriers for plutonium. Only a relatively small proportion of these, however, have the additional properties necessary for use in the present invention.

A suitable carrier for use in the present process comprises a compound containing at least one ion capable of a change in oxidation state or at least capable of easy destruction such that the other carrier ion may be transformed from a substantially insoluble compound into a compound which is substantially soluble in a solution in which the original carrier compound is substantially insoluble. A preferred carrier of this type comprises a compound Whose cation, or anion, or both may be oxidized to, form a compound which is soluble in a solution from which the carrier may be reprecipitated in its original state of oxidation, again carrying plutonium in its original oxidation state.

Ceric iodate is an example of a carrier whose cation and anion may be reduced simultaneously to form a soluble compound. Thus, cerous iodide may be maintained in solution while effecting a second precipitation of ceric iodate.

Uranous pyrophosphate is an example of a carrier havingy an oxidizable cation permitting its use in 'the present process. A uranous pyrophosphate precipitatecarrying tetravalent plutonium may be dissolved in hot nitric acid, thereby oxidizing the uranous ion to uranyl ion without` oxidizing the tetravalent plutonium. The resulting uranyl nitrate remains in solution even in the presence of the pyrophosphate ion. A second uranous pyrophosphate precipitation may then be made from this solution, using a smaller quantity of Uranous ion than in the original precipitate. The second precipitate will again 'carry the tetravalent plutonium with effective concentration by reduction of carrier-to-plutonium ratio.

Thorium oxalate is an example of a carrier having an oxidizable anion or anion `which may be decomposed permitting its use in the present process. oxalate precipitate carrying tetravalent plutonium may be dissolved in hot nitric acid, thereby oxidizing the oxalate ionto carbon dioxide, or decomposing this iony to carbon dioxide and carbon monoxide without oxidizing the tetravalent plutonium. A second thorium oxalate precipitation may then `be made `from the resulting tho rium nitrate solution. Other carboxylic acid salts of thorium, such as the citrate or tartrate, may be used in similar manner.

Uranous hypophosphate and uranous salts of carboxylic acid, such as uranous oxalate, are examples of carriers whose cations and anions may both be oxidized simultaneously to form soluble compounds. Thus, a uranous hypophosphate precipitate may be oxidized to uranyl orthophosphate, which may be maintained in solution while again precipitatinguranous hypophosphate. Similarly, auranous oxalate precipitate may be dissolved in hot nitric acid to form a solution of uranyl nitrate from which a second uranous oxalate precipitation may be made.

Also,` similar results can be achieved by the use of a suitable sulfide, for example, the suliides of thorium, uranium and the like.` When carriers of the above type are used they may be readily decomposed by the addition of an acid, such as nitric or hydrochloric acid.

' The specific carriers `referred to above are illustrative of the preferred classesof carriers for the present process. It `will be apparent to those skilled in the art that there are numerous other carriers which may be solubilized by a change in oxidation state of one or more ions. The choice of a desirable carrier for this process will depend upon the` oxidation-reduction potentials of the carrier ions and of the ion to be carried. If these potentials are too close together, it will be diliicult to effect selective oxidation or deuction of the carrier ion, or of the ion to be carried, without simultaneously changing the oxidation state of the other ion. For example, if tetravalent plutonium is to be carried by a carrier having a tetravalent cation, it is desirable either to effect highly selective oxidation of the carrier cation, or to oxidize both the plutonium and the carrier cation and `thereafter effect highly selective reduction of the hexavalent plutonium `to the tetravalent state.

Plutonium has a number of valence states, including el-3 +4, +5 and 5&6., In (L5-1.0 M. aqueous hydro- A thorium 'shlor'ie acid, the oxidation-reduction potentials are ofthe following magnitudes: Y

The lPu+ ion is 'generally very unstable, and disproportinates to Pui'4 and Pu+6. The Pu+4 ion is capable of disproportionating to the Pu'l'3 ion and the PuO2+2 ion, 'and in aqueous hydrochloric acid this disproportionation .may take place to a considerable extent.' The Pu+4 dis'- 1.o M. HC1 *0.97 v. 1.0 M. HC1-0.1 M. H3130., oso v.l 1.o M. HCl- 1.o M. HF 0.53 v.

Generally thefanions of slightly ionized acids tend to complex the Pu+4 ion to a inuch greater extent than the lanions of highly ionized acids. Thus, Pu+4 is only slightly complexed by C104, Cl", and NO3-g it is complexed to a much greater extent by S04-2; and it is very strongly Pof-3, F C21-1302, and C204- It may be seen from the above discussion that the +3 state (the trivalent ion), the +6 state (the uranyl ion), the stabilized +4 state (the plutonous ion), are the most important valence Vstates of plutonium. In `the present process it is preferred to carry plutonium in the +4 valence state, employing a carrier having a +4 cation. -If an oxidizable +4 carrier cation is to be 'chosen for this purpose, it should have an oxidation potiential substantially less positive than the oxidation potential of the Pu02+2 Pu+4 couple in the particular solution m'ployed. The following are representative potentials for" this couple;

1.o M. HC1l .+'1.0 v. 1.o M. HNO3 +1.1 v. 1.0 M. mso,L +13 v.

It is thus seen that if it is desired to oxidize the carrier cation in hydrochloric acid solution Without oxidizmg tetravalent'plutonium, the carrier cation should have an oxidation potentialsubstantially less positive than +.1.0 v.

vReference to' a table'of standard oxidation-reduction potentials such as that of Latimer and Hildebrand Vin the Handbook of Chemistry and Physics (Chemical Rubber Publishing Co., Cleveland, Ohio), will indicatewhich carrier cations are suitable for selective oxidation and will also indicate suitable selective oxidizing agents. "'`l`1us, the oxidation potential of the UO2+2 U+4 couple visy about +0.4 v., which indicates that tetravalent uranium may easily be oxidized selectively without oxidizing tetravalent plutonium.. An oxidizing agent such as the ferrie ion, with an oxidation potential substantially above +0.4 v. and substantially below +l.0 v., may be employed to effect selective oxidation of the U+4 ion.

It may also be seen that an alternative procedure can be employed. in the case of the U+4 carrier cation and the Pu+4 ion in hydrochloric acid solution. If an oxidizing-agent such as KZCrZO-l is used, the oxidation potential is suiciently high to oxidize Pu+4 to PuO2+2 simultaneously with the oxidation of U+4 to UO2+2. In this case, at the conclusion of. the oxidation the Pu02+2 may be selectively reduced by means of a reducing agent having a reduction potential substantially more negative than .-0.4 v. and substantially less negative than 1.0 v.

Hydrogen peroxide and the ferrous iron are suitable reducing agents for this purpose.

Since vthe oxidation potentlal of nitric acid is less positive than the oxidation potential of the PuO2+2-`- Pu'+4 couple-in nitric seid solution,v and substantially more posi.'-

`tive than the oxidation potential of the UOZ-F--SU'WE ',couple,l it may be s'een that nitric acid constitutes a se'- lective oxidizing agent for the U+4'ion. Thus, a precipitate of U+4 carrier andv its associated Pu+4 may be dissolved in hot nitric acid to form a solution of UO2+3 and Pu+4 ions, from which the U+4 carrier'rn'ay' again be precipitated in diminished quantity. t

Although the oxidation-reduction principles orf the present invention have been discussed above with particular reference to Ui'4 carriers, it is to be understood that the same general principles apply to the use of other carriers which may be solubilized by a change of oxidation state of cation, anion, or both cation and anion. v

Although the carriers in the present process may be employed as preformed nely divided solids, it is preferable to precipitate the carrier in situ since the latter yprocedure usually permits a lowercarrier ratio and results in more quantitative carrying of plutonium." The iirst precipitationcof any carrier in this process maybe effected in accordance with the techniques previously used in extraction or `decontamination processes. It is generally preferable to effect this precipitation by incorporating the carrier cation in the solution and then adding the anion while agitating the mixture. The techniques for eecting subsequent precipitations of the same` y'in 4subsequent precipitations of the carrier as was used in the initial precipitation. lf one lof the ions of the previous carrier remains in 'the new solution, however, it will usually be necessary to add only the other' ion to effect a reprecipitation. f

After the precipitation of the carrier in any step of the present rprocessit is desirable to digest the precipitate for a short time before separating it from the supernatant solution. hour to one hour or longer at the precipitation temperature will usually increasel the electiveness of the carrying. Care should be taken, however not to digest oxidizable precipitates for long periods `in solutions of oxidizing acids. Thus, although acid 'must be 'heated to ctie-ct rapid oxidation of a U+4 carrier, itwill oxidi'ze such a carrier slowly atv room temperature. A Ui'il carrier precipitated out of nitric acid should not, therefore, be digested for substantially more than anv hour before separation' from the supernatant solution.

'this invention will now be jfurther illustrated by :the following specific examplei Y f Y Example A dilute solution of plutonium, obtained by separating unreacted uranium and radioactive ssion products from a solution of neutron-bombarded uranium, isrvr concentrated by the procedure illustrated diagr'ammatically in the accompanying drawing.

The plutonium solution, resulting from dissolving a bismuth phosphate carrier precipitate and Vits `associated plutonium in 6-N. hydrochloric acid, comprises about 10,600 liters of solution containing aboutA 2,910 kilograms of bismuth phosphate and about grams of plutonium. This solution is introduced intothe irst precipitator and about 25 kilograms of Na4P2O6 is'then added as a saturated aqueous solution. Approximately 27.1 kilograms of U(SO4)2 in the form of a concentrated solution in dilute sulfuric acid is then introduced, while agitating the mixture. The resulting slurry is digested for one hourat 25 C., with continued agitation, and'is then transferred to the rst centrifuge fr separation of the precipitate.

Agitation of the slurry for one-half f sulting slurry is digested for three-fourths of an hourv at 25 C.,` with continued agitation, and is then transferred to a centrifuge for separation of the precipitate.

The hypophosphate precipitate is slurried from the cen- `trifuge to the second dissolver and oxidation reactor with sufiicient 6 N. nitric acid to form a solution ofapproximately 6 liters. The 'slurry is then heated for about onehalf hour to eect solution of the precipitate and oxidation of UPgO to (UO2)3(PO4)2. The nal solution thus obtained contains about 24.2 grams of Pu(NO3)4 per liter,

which is a suiciently high concentration to permit subsequent precipitation of an insolubleplutonium compound such as plutonium peroxide Without the necessity of coprecpitating a carrier.

The recovery of plutonium by the above procedure, without recycle of supernatant solutions, is about 89 percent of the quantity in the original solution. By recycling the supernatant solution from the last centri- -fuging operation to the preceding solution step, the recovery may be increased to about 98 percent.

It may be seen that in the above concentration procedure the volume of the final solution is only 0.06 percent of the volume of the original solution, -and the ratio of carrier to plutonium in the final solution is .only 0.02 percent of the ratio in the original solution.

It is to be understood of course, that the above example is merely illustrative and does not 'limit the Scope `of this invention. Other carriers which can be, solubilized by a change in oxidation state of one or more `ions may be substituted for the particular carrier employed in the example; and the procedure of the example may be `modiiied in numerous respects in accordance with the foregoing description. In general, it may `be said that the use of any equivalents or `modifications of procedure which would naturally occur to those skilled in the art is included in the scope of the present invention. Only such limitations should be imposed on the scope of this invention as are indicated in the appended claims.

What is claimed is:

1. A process of separating and concentrating tetravalent values of transuranic elements of an atomic number below 95 contained in an aqueous solution, comprising incorporatinga first carrier selected from the group consisting of uranium (IV) hypophosphate, uranium (IV) pyrophosphate, uranium (IV) oxalate, thorium oxalate, thorium citrate, thorium tartrate, thorium suliide and uranium (IV) sulde in said solution whereby said transuranic values are precipitated on said iirst carrier; separating the transuranic values-containing carrier from the solution; adding nitric acid to said carrier; heating the mixture of carrier and nitric acid whereby the carrier is converted to nitrate, the transuranic values are converted to nitrate of the tetravalent state and the nitrates are dissolved in the nitric acid; incorporating a second carrier in the nitric acid solution, said second carrier being chemically identical with said irst carrier and the amount of said second carrier being smaller than `the amount of said iirst carrier, whereby said transuranic `3L The process of claim 1 wherein said carriers are incorporated by separately adding to the solutionthe cationand the anion, each in the form of a water-soluble compound. L p f 4. lfhepr'ocess of claim 1` wherein the second' carrier is allowed to digestfor'from one-half to one hour prior to the separation from the aqueous solution.

5. A, process of separating and concentrating tetravalent plutonium values contained in an aqueous solution, ycomprising incorporating a rst carrier` selected from the group consisting of uranium (IV) hypophos' phate, uranium (IV) pyrophosphate, uranium p (1V) oxalate, thorium oxylate, thorium citrate, thorium tartrate, thorium suliide and uranium (1V) sulfide in said solution whereby said plutonium values are precipitated `on said iirst carrier; separating the plutonium-containing carrier `from the solution;` adding hot nitric acid to the carrier whereby the carrier is converted to nitratethe plutonium to ,plutonium (IV) nitrate and the nitrates are dissolved in the nitric acid; incorporating a second carrier in the nitric acid solution, said second carrier being chemically identical with said first ycarrier and the amount of said second carrier being smaller than the amount of said rst carrier, whereby said plutonium values are carried on said second carrier; and separating said second carrier'from the` solution. p

,6. A process of separating and concentrating plutonium Values present in the tetravalent state in an aqueous acid solution, comprising adding a first portion of an aqueous solution of sodium `hypophosphate; adding a rst portion of an aqueous solution of uranous sulfate; agtating the mixture obtained thereby at about 25 C. for approximately one hour whereby a precipitate of uranous hypophosphate carrying said plutonium values 4is formed; separating the precipitatefrom the. solution; adding a first portion of nitric acid to the precipitate;

heating the mixture of precipitate and nitric acid to about 70 C. for about 30 minutes whereby the hypophosphate is oxidized to the orthophosphate and the uranyl and plutonium (IV) phosphates are dissolved in thelnitric acid; adding a second portion of an aqueous sodium hypophosphate solution and a second portion of an aqueous uranous sulfate solution to the nitric acid solution formed, said second portions being considerably smaller than said tirst portions previously added, whereby a carrier precipitate forms which contains the plutonium values; digesting the mixture with the carrier precipitate at about 25 C. for about 45 minutes; separating the carrier precipitate from the solution; adding a second portion of nitric acid to the carrier precipitate, said second portion being substantially smaller than said iirst portion of nitric acid previously added; heating the nitric-acidcontaining mixture for about 30 minutes whereby uranyl phosphate and plutonium (IV) phosphate are formed and dissolved and a comparatively concentrated` plutonium solution is formed; and precipitating the plutonium values from the concentrated solution.

References Cited in the iile of this patent Smyth: A General Account of the Development of Methods of Using Atomic Energy for Military Purposes Under the Auspices of the U.S. Government, i940- 1945, pp. 99, (August 1945), U.S. Government Printing Otiice.

'N-2205, Summary of Plutonium Carrying Agents, U.S. Atomic Energy Commission Document dated January 16, 1946, declassied Nov. 22, 1957, pp. 20-24, 32, 41, 42. (The information disclosed on these pages was taken from reports CN-9l4, September l, 1943; CN-979, September 30, 1943; CN-l241, January 5, 1944, and these dates are relied on to show prior use or knowledge of the invention under secrecy within the Atomic Energy Program of the U.S.; see Sec. 155, Atomic Energy Act of 1954.) V 

1. A PROCESS OF SEPARATING AND CONCENTRATING TETRAVALENT VALUES OF TRANSURANIC ELEMENTS OF AN ATOMIC NUMBER BELOW 95 CONTAINED IN AN AQUEOUS SOLUTION, COMPRISING INCORPORATING A FIRST CARRIER SELECTED FROM THE GROUP CONSISTING OF URANIUM (IV) HYPOPHOSPHATE, URANIUM (IV) PYROPHOSPHATE, URANIUM (IV) OXALATE, THORIUM OXALATE, THROIUM CITRATE, THROUM TERTRATE, THRORIUM SULFIDE AND URANIUM (IV) SULFIDE IN SAID SOLUTION WHEREBY SAID TRANSURANIC VALUES ARE PRECIPITATED ON SAID FIRST CANRIER; SEPARATING THE TRANSURANIC VALUES-CONTAINING CARRIER FROM THE SOLUTION; ADDING NITRIC ACID TO SAID CARRIER; HEATING THE MIXTURE OF CARRIER AND NITRIC ACID WHEREBY THE NCARRIER IS COVERTED TO NITRATE, THE TRANSURANIC VALUES ARE CONVERTED TO NITRATE OF THE TETRACALENT STATE AND THE NITRATES ARE DISSOLVED IN THE NITRIC ACID; INCORPORATING A SECOND CARRIER IN THE NITRIC SOLUTION, SAID SECOND CARRIER BEING CHEMICALLY IDENTICAL WITH SAID FIRST CARRIER AND THE AMOUNT OF SAID SECOND CARRIER BEING SMALLER THAN THE AMOUNT OF SAID FIRST CARRIER, WHEREBY SAID TRANSURANIC VALUES ARE CARRIED ON SAID SECOND CARRIER; AND SEPARATING SAID SECOND CARRIER FROM THE SOLUTION. 